A model for pellet-cladding interaction (PCI) fracture of light-water reactor (LWR) fuel rods is presented, the basis of which is that Zircaloy cladding fails by iodine stress corrosion cracking (SCC). Laboratory data on iodine SCC of irradiated Zircaloy provide the primary input to the model, but unirradiated Zircaloy SCC data and theoretical analyses are utilized to broaden the regime of validity to encompass the various power reactor observations.

The PCI model postulates a threshold stress below which cladding cracks are unable to form. The cladding hoop stress must exceed this threshold for a finite time at any particular fission product iodine availability for the crack(s) to form. Crack formation is the controlling event in cladding failure due to SCC, and therefore, “remedies” must address the situation at the cladding inner surface.